Abstract: A method for determining tube support plate blockage of a steam generator includes the following steps: measuring at least five different eddy current values per tube support intersection; calculating a nominal clean fit radius of flow hole; determining a center signal response; converting the center signal response to a deposit thickness; determine an edge reduction; converting the edge reduction to an edge thickness; calculating the resulting flow hole radius; verifying the reasonableness of the resulting flow hole radius; and determining a virtual calibration range.

Abstract: A method for chemical decontamination of an oxide-coated surface of a metal structural part or of a system in a nuclear power plant with several cleaning cycles, involves oxidation steps, in which the oxide layer is treated with an aqueous solution containing an oxidation agent, and a subsequent decontamination step, in which the oxide layer is treated with an aqueous solution of an acid. At least one oxidation step is carried out in an acid solution and at least one oxidation step in an alkaline solution.

Abstract: A protective cover is provided for a ventilation opening in a plant space of a nuclear-engineering plant. The cover is configured, with a construction that is simple, and with simple manufacture and mounting, for a sufficient air-flow cross section and also for protection against ingress of destructive projectiles into the building interior. The protective cover has profile elements, which are arranged in a mounting plane in a rectangular frame, are aligned in a longitudinal direction such that they are parallel to one another, and are shaped in the same way and orientated in the same way. The relevant profile element has an L-shaped or V-shaped cross section with two limbs that meet in a peak, with the peaks pointing in the transverse direction. A spacing between profile elements that immediately follow each other is chosen such that there is no rectilinear passage between them.

Abstract: An assembly is provided that includes at least one first conduit and at least one second conduit connected together through a connecting device. The connecting device includes an exteriorly threaded tube; a connecting part attached to one end of the tube and including an orifice; a holding nut screwed onto the tube and axially maintaining the connecting part relative to a tube; and a retaining member attached on the holding nut and engaging with the first conduit so as to axially retain the first conduit fitted into the orifice, the retaining member cooperating with the first conduit so as to oppose the unscrewing of the holding nut.

Abstract: A method for measuring the uranium concentration of an aqueous solution including the following successive steps: a) electrochemical reduction towards valence IV, of the uranium present in the aqueous solution with a valence greater than IV, this reduction being implemented at pH<2 and by passing an electrical current in the solution; b) measurement of the absorbance of the solution obtained on completion of step a) at a chosen wavelength between 640 and 660 nm, and preferably 652 nm; and c) determination of the uranium concentration of the aqueous solution by deduction of the uranium concentration of valence (IV) present in the aqueous solution obtained on completion of step a) from measurement of the absorbance obtained in step b).

Abstract: An improved melting furnace including a crucible and a plurality of parallel conductors of identical height surrounding the crucible having at least one descending portion (9) and one ascending portion (10). The benefit from this arrangement is that the conductors all have a portion located at each heating height which guarantees density uniformity of the currents flowing in the conductors even if the load of the crucible has superimposed regions for which the electrical resistivity is different.

Abstract: A nuclear engineering plant has a containment, whose interior chamber is subdivided by a wall into a systems chamber and an operating chamber which is accessible during normal operation. The containment ensures a particularly high operational reliability, in particular also in incident situations, in which hydrogen is released in the systems chamber. For this purpose, a number of overflow openings are provided in the partition wall, the respective overflow opening is closed by a closure element of a closure apparatus which opens automatically when a trigger condition associated with the respective overflow opening is reached. Closure apparatuses are provided which open both as a function of pressure and independently of pressure. The closure apparatus furthermore has a closure element containing a bursting film or a bursting diaphragm. The closure apparatus is configured such that it frees the overflow opening automatically when a predetermined environment-side trigger temperature is reached.

Abstract: A fuel element for a nuclear reactor has a fuel rod bundle, at least one spacer with cells defined by at least one web section made from a first material and several guide tubes each running through a cell and axially fixed thereto made from a second material. The first and second materials have differing thermal expansion coefficients. The connection between the guide tube and the spacer is embodied as follows: first and second projections are directly or indirectly fixed to the guide tube. The first projections are disposed in a first axial position and the second projections are arranged at a second axial position and the projections each engage in an opening through the web section to give an axially-acting undercut.

Abstract: A method for controlling the positions of a plurality of nuclear fuel assemblies (1) relative to an upper core plate (3) in a nuclear reactor core, the method including the following steps: choosing a reference point (13) in internals or in a reactor vessel; determining the positions of S-shaped holes of the nuclear fuel assemblies (1) relative to the reference point (13), each S-shaped hole being intended to cooperate with a corresponding centering pin of the upper core plate (3); acquiring the positions of the centering pins of the upper core plate (3) relative to the reference point (13); and comparing the positions of the S-shaped holes and the positions of the pins and deducing therefrom whether the nuclear fuel assemblies (1) are correctly positioned relative to the upper core plate (3).

Abstract: An electrical circuit breaker, in particular a high-voltage circuit breaker filled with insulating gas, is described. The circuit breaker is furnished with a first contact, in particular, a contact pin and a second contact, in particular, a tulip contact that are movable in opposite directions. The circuit breaker is furnished with a drive mechanism that is coupled to the second contact. The circuit breaker is furnished with a reversing gear that produces a coupling between the second and the first contact. A first indicator element is provided that is associated with the reversing gear.

Abstract: A component for conducting or receiving a fluid, in particular a component of a fluid-conducting line system of an industrial plant, especially of a line system of a tertiary cooling circuit of a nuclear power plant, includes a wall having a supporting structure made of a glass-fiber-reinforced plastic. Electrically insulating inner and outer protective layers are disposed on respective inner and outer surfaces of the supporting structure. An electrically conductive inner intermediate layer lies between the inner protective layer and the supporting structure and is provided with an electrical terminal. An electrically conductive outer intermediate layer lies between the outer protective layer and the supporting structure, is provided with an electrical terminal and is electrically insulated from the inner intermediate layer. A method for testing the component is also provided.

Abstract: A method and a device depressurize a nuclear power plant. A depressurization flow is conducted out of a containment shell into the atmosphere via a depressurization line having a filter system. The filter system contains a filter chamber having an inlet, an outlet, and a sorbent filter. The depressurization flow is first conducted in a high-pressure section, then is depressurized by expansion at a throttle device, then conducted through the filter chamber having the sorbent filter, and finally blown out. To enable an effective retention of activity carriers contained in the depressurization flow, including organic compounds containing iodine, the depressurization flow depressurized by the throttle device is conducted through a superheating section before the depressurization flow enters the filter chamber, in which superheating section the depressurization flow is heated from the not yet depressurized depressurization flow to a temperature that is at least 10 ° C. above the dew point temperature.

Abstract: In a process for estimating the concentration (C) of a chemical element in the primary coolant of a nuclear reactor, a dilution solution or a concentrated solution of said chemical element in a predetermined concentration (C*) is injected into the primary coolant within the reactor, and the reactor includes a sensor capable of measuring a quantity (Cm) representing the concentration of said chemical element. The process is an iterative process in which repeatedly in each time step k: a stage of acquiring quantities (qdk) and (qck) representing the injected flows of dilution solution and concentrated solution in step k, and a quantity (Cmk) representing the concentration measured by the sensor; a stage of calculating an estimated value (Cek+1) of the concentration of said chemical element in the primary coolant in step k+1 on the basis of representative quantities (qdk, qck, Cmk) acquired in stage k.

Abstract: The invention relates to a method of determining at least one technological uncertainty factor in respect of nuclear fuel elements (23) as a function of variations in the production parameters of the elements (23) in relation to nominal values. The inventive method comprises a step involving the use, for at least one production parameter, of a collective variation in said parameter in relation to the nominal value within a batch of produced elements (23). The invention can be used, for example, to design, produce and check pellets for light water reactors.

Abstract: A device for the dry handling of nuclear fuel assemblies is provided. The device includes a transfer basket which can be connected to a lifter and which includes a gripper for gripping the fuel assembly to be transferred, the gripper being supported by a lift built into the basket; and an indexing table which can be placed on a cask and which comprises a positioner for positioning the basket over a slot in the cask.

Abstract: An assembly is provided including a mobile structure including a main pipe equipped with a first end intended to be connected to a water supply and a second end intended to be connected to a circuit connected to the primary circuit of the reactor, and including between these two ends in the direction of circulation of the water, a pump, a water heating device, an injector for continuously injecting the powdered neutron-absorbing element into the water of the main pipe, a first mixer for mixing and dissolving powder with water and a controller driving and controlling the flow rate of the water and the flow rate of the powder injected.

Abstract: Equipment to transfer powder from a drum into a tank comprising at least one handling device and one hopper assembly (T) connected to a tank (4). The handling device comprises a hollow cylindrical body intended to surround a drum (2) and to immobilise it to enable it to be turned upside-down, where the drum (2) is open and the powder is contained in at least one sealed bag. The hopper assembly (T) comprises a hopper (60) and a sealed enclosure (62) installed on the upstream end of the hopper, where said enclosure (62) comprises an aperture to enable the hollow cylindrical body loaded with the drum (2) and of the sack of powder to enter, and then, after removal of the hollow cylindrical body loaded with the empty drum (2), to be closed, and to allow the bag to be opened to allow the powder to flow into the hopper (60).

Abstract: A nuclear fuel assembly is provided for a boiling water reactor. The nuclear fuel assembly includes a base, a head, and a bundle of full length fuel rods and partial length fuel rods, said bundle extending upwardly and longitudinally from the base to the head. The nuclear fuel assembly includes at least one clamp for longitudinally retaining a lower plug of a partial length fuel rod with respect to the base. The clamp is an additional part fitted to the base, the clamp is at least partially received in a housing provided in the base, and the clamp is assembled to the base by mechanical engagement of complementary assemblies.

Abstract: A method takes a sample of a deposit on a secondary side of a pipe base plate of a steam generator of a nuclear power plant, In the method a steam generator pipe is removed from the pipe base plate to expose a hot pipe bore penetrating the pipe base plate. A removal tool of a device for taking the sample is introduced into the hot pipe bore by the primary side of the pipe base plate which is opposite the secondary side. A part of the deposit is removed by the removal tool as the sample. The sample is transported and removed from the steam generator. The removal tool is removed from the steam generator. The device contains the removal tool which can be introduced by a primary side of the pipe base plate in an exposed hot pipe bore penetrating the pipe base plate for removing the sample.

Abstract: A heat treatment method for desensitizing a nickel-based alloy with respect to environmentally-assisted cracking, the alloy having the following composition in percentages by weight: C?0.10%; Mn?0.5%; Si?0.5%; P?0.015%; S?0.015%; Ni?40%; Cr=12%-40%; Co?10%; Al?5%; Mo=0.1%-15%; Ti?5%; B?0.01%; Cu?5%; W=0.1%-15%; Nb=0-10%; Ta?10%; the balance being Fe, and inevitable impurities that result from processing, characterized in that the alloy is held at 950° C.-1160° C. in an atmosphere of pure hydrogen or containing at least 100 ppm of hydrogen mixed with an inert gas. A part made of a nickel-based alloy having the composition and that has been subjected to the heat treatment.